05000328/LER-2005-001

LER-2005-001, Unit 2 Reactor Trip Following Closure of Main Feedwater Upon Inadvertent Opening of Control Breakers

Docket Number
Event date:	02-23-2005
Report date:	04-25-2005
Reporting criterion:	10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function
3282005001R00 - NRC Website

I. PLANT CONDITION(S)

Unit 2 was in Power Operation (Mode 1) at approximately 100% thermal power.

II. DESCRIPTION OF EVENT

A. Event:

On February 23, 2005, at approximately 2030 Eastern standard time (EST), Maintenance personnel were performing pre-refueling outage maintenance activities on a 125-Volt (V) direct current (dc) vital battery board [EllS Code BYBD] molded case circuit breaker [El IS Code BKR]. After removing the breaker from the panel, Maintenance personnel began installation of a battery board panel cover to its normal position. While aligning the cover to insert the holding bolts, the cover shifted downward, opening two adjacent breakers. One of the adjacent breakers controlled a fuse [El IS Code FU] column providing power to all four main feedwater regulation valves [EllS Code FCV], which subsequently closed on loss of control power. A reactor trip was initiated by the reactor protection system [EllS Code JC] on low-low steam generator (SG) [EllS Code SG] level. As designed, auxiliary feedwater (AFW) system [El IS Code BA] was automatically initiated on SG low-low level.

B. Inoperable Structures, Components, or Systems that Contributed to the Event:

None.

C. Dates and Approximate Times of Major Occurrences:

February 23,� Maintenance personnel began physical breaker work 2005� activities in the dc vital battery board panel.

at -2030 EST February 23,� Maintenance personnel began restoration of the dc 2005 at� vital battery board panel.

-2058 EST February 23,� Direct current vital battery board panel cover shifts 2005� downward opening two adjacent breakers, resulting at -2106 EST� in closure of the main feedwater regulation valves; subsequently, the unit received a low-low generator level signal initiating a reactor trip.

D. Other Systems or Secondary Functions Affected:

None.

E. Method of Discovery:

The indication of an event began with an annunciation for high feedwater heater pressure [EllS Code PI] on the main control panel. Shortly thereafter, multiple alarms were received, including low SG level in the main control room.

Subsequently, a reactor trip was annunciated on the main control panel [EllS Code CBD].

F. Operator Actions:

Control room operators responded to the event in accordance with plant procedures. The first alarm received was for high feedwater heater pressure, which operators responded to by placing the main feed pump master controller [EllS Code SCO] in manual. Subsequently, multiple alarms were received including low SG level. Operators found the main feedwater regulation valves closed and non-responsive when placing them in manual control in response to the low SG alarm. Operators determined a manual reactor trip was necessary; however, an automatic reactor trip occurred as a result of low-low SG level before the manual reactor trip was carried out. Following the automatic reactor trip, operators took actions necessary to stabilize the unit, and maintained the unit in Hot Standby, Mode 3.

G. Safety System Responses:

The plant responded to the reactor trip as designed, including automatic start of all trains of the AFW system. Operators reduced the flow of AFW following the reactor trip to mitigate a decrease in reactor coolant system (RCS) [El IS Code AB] average temperature. Operators noted the Rod Position Indication (RPI) System [EllS Code BWR] indicated Shutdown Bank D Rod Cluster Control Assembly (RCCA) approximately 50 steps withdrawn during the reactor trip and later indicated zero steps. However, the RPI also indicated the RCCA fully inserted by the rod bottom indication light. A review of the integrated computer system [El IS Code ID] verified the RCCA did not stop at approximately 50 steps withdrawn as indicated by the RPI system. This abnormality has been documented in the past as a result of static electricity or the meter [El IS Code MTR] pointer rubbing on the meter scale of the RPI.

III.0 CAUSE OF THE EVENT

A. Immediate Cause:

The immediate cause of the unit trip was the initiation of an automatic reactor trip signal, on low-low level in the Loop 2 SG, when the main feedwater pumps were unable to provide feedwater flow to maintain SG level as a result of the closed feedwater regulation valves.

B. Root Cause:

The root cause of the inadvertent closure of the trip sensitive breakers was determined to be insensitivity to trip risks.

C. Contributing Factor:

Contributing to the event was a lack of risk sensitivity controls for determining at what plant operational mode this maintenance activity should be performed. Site barriers, including work week senior reactor operator review, critical evolutions committee review meeting and the maintenance foreman provided pre-job brief failed to understand the scope of the risk. The first of the two pre-job briefs had not addressed the risk to generation; however, once recognized in the second briefing, no escalation to management was taken nor was the activity stopped or mitigative action taken to reduce the risk. Also, placards warning of unit trip hazards were not in place to enhance sensitivity of the work environment and the work instruction did not contain cautions warning of unit trip hazards.

IV.0 ANALYSIS OF THE EVENT This event is most similar to and bounded by the analyzed event of Loss of Normal Feedwater in the Final Safety Analysis Report, with all plant safety systems operating as designed during and following the reactor trip. Prior to the event, the plant was in Mode 1 with the following conditions: RCS pressure was approximately 2235 pound per square inch (psig) with an average temperature near program value of 578.2 degrees Fahrenheit (°F); pressurizer [El IS Code PZR] level was on program at approximately 60 percent (%); secondary side steam pressure was approximately 845 psig with normal main feedwater supply and nominal full power steam flow; and SG levels were at 44% narrow range (NR). Following the automatic reactor trip, the RCS pressure increased to a maximum of 2285 psig upon loss of normal heat removal, subsequently reducing to a minimum of 2070 psig as a result of RCS temperature decrease and coolant volume shrinkage. The loss of nuclear heat generation and the introduction of cold AFW resulted in a rapid decrease in RCS temperature to below 540°F for approximately 4 minutes, and reached a minimum of 537°F. Operators took action to reduce RCS cooldown, and before procedural steps to emergency borate were carried out, RCS temperature had increased above 540°F no longer requiring emergency boration. RCS temperature was restored to 550°F approximately 30 minutes after the reactor trip.

Pressurizer level decreased to 22% and stabilized near 30% following the event. Steam pressure increased to approximately 1060 psig when the turbine stop valves closed. The atmospheric relief valves (ARVs) [EllS Code RV] on SG Nos. 1, 2, & 4 opened as designed. The ARV on SG No. 3 did not open during the transient and was investigated and repaired under a work document. Steam generators' safety relief valves did not open and secondary side pressure limits were not exceeded. Additionally, steam dumps to the condenser [EllS Code COND] operated as expected and remained available.

AFW system actuated as designed on the SG low-low level signal, recovering the SG water levels following the reactor trip.

V. ASSESSMENT OF SAFETY CONSEQUENCES

Based on the above "Analysis of The Event," this event did not adversely affect the health and safety of plant personnel or the general public.

VI. CORRECTIVE ACTIONS

A. Immediate Corrective Actions:

Immediate action was taken to require the use of guide rods (i.e., positive position control) for work on similar breakers. Additionally, the maintenance procedure in use prior to the event has been revised to include the use of guide rods for all 120-V alternate current and 125-V dc boards. Pre-outage work was stopped until a procedure was implemented to provide an additional review of nuclear safety and generation risks against site work. Plant management reviewed pre-outage work for release to work until the additional procedure was implemented, following the plant event. Warning placards have been installed on the battery board room doors to enhance the level of sensitivity of the work environment. Additionally, the craftsmen involved in the maintenance activity have been restricted from generation risk maintenance activities until they demonstrated the correct work standards.

B. Corrective Actions to Prevent Recurrence:

SQN has implemented a procedure that provides additional barriers for review and control of site work activities to minimize and manage potential nuclear safety and generation risk above and beyond that specified in the TVA nuclear online and outage work management procedures. The procedure includes the following attributes: specific questioning of the work activity relative to it causing a plant transient; questions concerning reactivity management, containment integrity, and electrical systems; contingency requirement if the proximity of other equipment poses a risk; and a graduated level of reviews up to a review by senior management.

The maintenance procedure for work on molded case breakers has been revised to include positive position control of all panel covers.

A. Failed Components:

None B. Previous LERs on Similar Events:

A review of previous reportable events for the past three years did not identify any previous similar events.

C. Additional Information:

None D. Safety System Functional Failure:

This event did not result in a safety system functional failure in accordance with 10 CFR 50.73(a)(2)(v).

E. Loss of Normal Heat Removal Consideration This event did result in a loss of normal heat removal because the closure of the main feedwater regulation valves precipitated the reactor trip.

VIII. COMMITMENTS

None.